ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

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Document Title:

Idaho National Engineering and 

Environmental Laboratory – Occupational 

External Dosimetry

Document Number: ORAUT-TKBS-0007-6 

Revision: 01 

Effective Date: 11/07/2006

Type of Document: TBD

Supersedes: Revision 00

Subject Experts: Norman D. Rohrig and Steven L. Bump

Approval: Clark Barton Signature on File for Approval Date: 11/02/2006

Daniel Mantooth, Document Owner

Approval: Signature on File Approval Date: 10/26/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/25/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 11/03/2006

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 11/07/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 2 of 51

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

04/06/2004 00 New technical basis document for the Idaho National Engineering 

and Environmental Laboratory (INEEL) – Occupational External 

Dosimetry. First approved issue. Initiated by Norman D. Rohrig.

11/07/2006 01 Document revised as a result of biennial review. Approved Revision 

01. Incorporates formal internal review comments. Revision 

constitutes a total rewrite of document. Adds Section 6.5.1 to 

incorporate comments from the Worker Outreach meeting with the 

PACE Local 8-0652 on April 28, 2004. Incorporates formal NIOSH 

review comments. This revision results in no change to the assigned 

dose and no PER is required. Training required: As determined by 

the Task Manager. Initiated by Norman D. Rohrig.
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TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

5

6.1 Introduction ..............................................................................................................................7

6.1.1 Purpose........................................................................................................................8

6.1.2 Scope...........................................................................................................................8

6.2 Basis of Comparison................................................................................................................8

6.3 Dose Reconstruction Parameters........................................................................................... 11

6.3.1 Site Administrative Practices ......................................................................................11

6.3.2 Personnel Monitoring Systems ...................................................................................16

6.3.2.1 Initial Film Badge ......................................................................................... 16

6.3.2.2 Multiple-Filter NRTS Film Badge.................................................................. 17

6.3.2.3 Original Lithium Fluoride Teflon TLD System............................................... 18

6.3.2.4 Automatic Thermo Luminescent Analyzer System ....................................... 19

6.3.2.5 Harshaw Two-Chip TLD System.................................................................. 19

6.3.2.6 Panasonic Four-Chip System ...................................................................... 19

6.3.2.7 Nuclear Track Emulsion, Type A for Neutrons ............................................. 20

6.3.2.8 Neutron Albedo Dosimetry ........................................................................... 21

6.3.3 Calibration..................................................................................................................21

6.3.3.1 Beta-Gamma Radiation................................................................................ 21

6.3.3.2 Neutron Calibration...................................................................................... 22

6.3.4 Workplace Radiation Fields........................................................................................24

6.3.4.1 Gamma Radiation........................................................................................ 24

6.3.4.2 Beta Radiation ............................................................................................. 27

6.3.4.3 Neutron Radiation........................................................................................ 29

6.3.4.4 Materials Test Reactor Neutron Radiation ................................................... 32

6.3.4.5 Test Area North Fuel Storage Casks............................................................ 34

6.3.4.6 Typical Workplace Neutron Dosimeter Hp(10) Performance......................... 34

6.4 Adjustments to Recorded Dose.............................................................................................. 34

6.4.1 Neutron Weighting Factor...........................................................................................34

6.5 Missed Dose ..........................................................................................................................36

6.5.1 Dosimeter Not Worn...................................................................................................36

6.5.2 Photon Missed Dose ..................................................................................................36

6.5.3 Missed Beta Dose ......................................................................................................36

6.5.4 Missed Neutron Dose................................................................................................. 38

6.5.4.1 Before October 1976.................................................................................... 39

6.5.4.2 After September 1976.................................................................................. 39

6.6 Organ Dose............................................................................................................................39

6.7 Uncertainty.............................................................................................................................40

References ........................................................................................................................................41

Glossary ............................................................................................................................................49
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LIST OF TABLES

TABLE TITLE

6-1 Area codes that could be in claimant dose files......................................................................

PAGE

13

6-2 Reasons Codes .....................................................................................................................14

6-3 Irregularity codes ................................................................................................................... 14

6-4 Column 20 codes ................................................................................................................... 14

6-5 1981 FNCFs...........................................................................................................................21

6-6 Laboratory sources of uncertainty for beta/photon dosimeter calibration 

parameters.............................................................................................................................23

6-7 Common sources of laboratory bias in the calibration parameters for neutron 

dosimeters .............................................................................................................................24

6-8 Penetrating photon dose (rem) percentiles by year for monitored workers ............................. 25

6-9 Selection of IREP beta and photon energies for INEEL facilities ............................................ 27

6-10 Beta dosimeter thicknesses and associated underreporting................................................... 29

6-11 Facilities and time periods for neutron exposure .................................................................... 30

6-12 Typical workplace neutron dosimeter Hp(10) performance ..................................................... 35

6-13 Calculated and recommended dose equivalent fractions and quality factor 

corrections .............................................................................................................................36

6-14 Beta/photon dosimeter period of use, type, MRL, exchange frequency, and 

potential annual missed dose................................................................................................. 37

6-15 Neutron dosimeter type, period of use, exchange frequency, laboratory 

minimum detectable limit, and maximum annual missed dose ............................................... 38

6-16 Recommended IREP neutron energy fractions and correction factors.................................... 39

LIST OF FIGURES

FIGURE TITLE

6-1 Individual dose reporting form in use until 1958 .....................................................................

PAGE

12

6-2 Film report form used in 1958 ................................................................................................ 12

6-3 Monthly badge reporting form from May 1959........................................................................ 13

6-4 Badge pull results from January 1961 for work in recovery from the SL-1 accident................ 15

6-5 Special badge report associated with a high beta reading listed in Figure 6-4 ........................ 15

6-6 Personnel exposure questionnaire partially completed for a hypothetical case ...................... 16

6-7 Response of DuPont 508 film with various filters to 140 mrem uranium beta 

and 100 mR of different energy photon irradiation.................................................................. 18

6-8 Probability density of neutron spectrum from a 241Am-Be (α,n) source ................................... 23

6-9 Collective dose for INEL personnel from 1951-1990 .............................................................. 26

6-10 Gamma and beta radiation field characterization ................................................................... 26

6-11 Distribution of beta ranges ..................................................................................................... 28

6-12 Distribution of reportable neutron dose for the first 9 months of 1995..................................... 30

6-13 Neutron radiation field characterization .................................................................................. 31

6-14 Neutron spectra simulating INEEL facilities............................................................................ 31

6-15 Sample MTR spectra from Hankins Bonner measurements ................................................... 32

6-16 MTR neutron field components .............................................................................................. 32

6-17 Correlation of fast neutron dose equivalent to gamma dose at MTR ...................................... 33
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANL-W Argonne National Laboratory-West

ANSI American National Standards Institute

AP anterior-posterior

ARA Army Reactor Area/Auxiliary Reactor Area

CFA Central Facilities Area

Ci curie

cm centimeter

CPP Chemical Processing Plant

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EBR Experimental Breeder Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EPRI Electric Power Research Institute

ERDA U.S. Energy Research and Development Administration 

FNCF facility neutron correction factor

g gram

H*

(10) ambient dose equivalent 

Hp(10) personal dose equivalent at 10 mm depth in tissue

Hp(d) personal dose equivalent

Hp,slab(d) personal dose equivalent (slab phantom)

hr hour

ICPP Idaho Chemical Processing Plant (formerly CPP and now INTEC)

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

in. inch

INEEL Idaho National Engineering and Environmental Laboratory

INEL Idaho National Engineering Laboratory

INTEC Idaho Nuclear Technology and Engineering Center (formerly ICCP and CPP)

IREP Interactive RadioEpidemiological Program 

keV kiloelectron-volt, 1,000 electron-volts

LET linear energy transfer

LOFT Loss of Fluid Test

LPTF Low Power Test Facility

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mm millimeter

mR milliroentgen

mrad millirad
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mrem millirem

MRL minimum reporting level

MTR Materials Test Reactor

n neutron

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurement

NIOSH National Institute for Occupational Safety and Health

NRTS National Reactor Testing Station

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

PBF Power Burst Facility

POC probability of causation

R roentgen

RBE relative biological effectiveness

RESL Radiological and Environmental Services Laboratory

RWMC Radioactive Waste Management Complex

SMC Specific Manufacturing Capability

SPERT Special Power Excursion Reactor Test

TAN Test Area North

TLD thermoluminescent dosimeter

TRA Test Reactor Area

TREAT Transient Reactor Test

TSF Tower Shielding Facility

U.S.C. United States Code

WERF Waste Experimental Reduction Facility

wk week

yr year

Z atomic number

ZPPR Zero Power Plutonium Reactor, Zero Power Physics Reactor

α alpha particle

§ section
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6.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist the NIOSH staff in the completion of the individual work required for each dose 

reconstruction. 

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work. 

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
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• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

6.1.1

This document describes the processes and results for measuring external doses at the Idaho 

National Engineering and Environmental Laboratory (INEEL) and its predecessor organizations [the 

National Reactor Testing Station (NRTS) and the Idaho National Engineering Laboratory (INEL)]. In 

2005, DOE shortened the name to the Idaho National Laboratory, but this document uses INEEL for 

consistency. 

Purpose

6.1.2

From the start of operations in 1951 (when it was NRTS), a branch of the U.S. Atomic Energy 

Commission (AEC) Idaho Operations Office provided external dosimetry resources and services at 

INEEL until 1989, when DOE transferred that responsibility to the prime operating contractor. Despite 

the fact that INEEL had several contractors at a time and that contractors changed often, the external 

dosimetry process has remained under technical management of a single organization with 

responsibilities for dosimetry development, operational dosimetry, and radiological records; thus 

providing a stable external dosimetry system. 

Scope

6.2 BASIS OF COMPARISON

The Interactive RadioEpidemiological Program (IREP) calculates the probability of cancer induction in 

an organ from the external equivalent dose and internal dose received by that organ. Appendix B of 

the External Dose Reconstruction Implementation Guideline (NIOSH 2002) provides conversions from 

four photon dose quantities [deep dose equivalent Hp(10), ambient dose equivalent H*

(10), exposure 

X, and air kerma Ka] and three neutron quantities [fluence φ, ambient dose equivalent H*

(10), and 

deep dose equivalent, Hp,slab(10)] to the organ doses. Over the years, the National Council on 

Radiation Protection and Measurements (NCRP), the International Commission on Radiological 

Protection (ICRP), and their predecessor organizations have developed the definitions of dosimetry 

parameters. Dose parameters measured by the INEEL dosimetry system have received further 

definition. INEEL has reported doses as penetrating and nonpenetrating. The penetrating dose 

corresponds to the deep dose equivalent, and the sum of the nonpenetrating and penetrating doses 

corresponds to the shallow dose equivalent. 

Horan and Braun (1993), Attix and Roesch (1968), and Meinhold (1975) discuss the history of 

radiation protection requirements from the 1930s. On September 29-30, 1949, representatives of the 

United Kingdom, Canada, and the United States met in Chalk River, Canada at a Permissible Doses 

Conference (Warren et al. 1949) and picked the blood forming organs as critical for hard x-rays or 

gamma rays at a depth of 5 cm. Skin is the critical organ for soft x-rays or beta radiation. Dosage 

was to be reported in roentgen equivalent physical (rep) units equivalent to 93 ergs absorbed per gm

of tissue. The maximum permissible dose to the bone marrow was 0.3 rep/wk.

In 1949, the newly formed National Committee on Radiation Protection (now the National Council on 

Radiation Protection and Measurement or NCRP) issued NCRP Report 7 through the National Bureau 

of Standards as Handbook 42 (NBS 1949, p. 6), which recommended a permissible dose of 0.3 R/wk

(15 R/yr) for occupational workers. The term dose was undefined. A roentgen was defined as the 

quantity or dose of X-rays such that the associated ionization per 0.001293 g of air (1 cm3 at standard 

temperature and pressure) produces 1 electrostatic unit of charge of either sign. An April 1952 site 

manual stated the limit as, “0.3 rep/wk at an effective depth in soft tissue of 5 cm, assumed to be the 
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depth of the blood forming organs” (ACC 1952, p. IV: 1-1). The manual does not mention a quarterly 

or an annual limit.

In 1953, the International Commission on Radiation Units and Measurements (ICRU) established a 

new quantity, absorbed dose, which is the energy deposited in material per unit mass by radiation 

using the rad (radiation absorbed dose), which was defined as 100 ergs/g (ICRU 1954). It specified 

the term exposure dose, later to become exposure, for the ionization capability in air for X- and 

gamma rays. In 1956, the ICRU defined the term relative biological effectiveness (RBE) dose and the 

quantity rem (roentgen equivalent in man) and the concept of adding all types of external doses 

together (ICRU 1956). 

In 1957, the NCRP introduced an age prorating formula for the maximum allowable dose of 5 rem per 

year above age 18 (5 rem × [age (yr) -18]) (NBS 1958). This introduced 5 rem as an average annual 

dose but deemphasized that value as a limit. The AEC issued AEC Manual Chapter 0524 

“Permissible Levels of Radiation Exposure” on January 9, 1958, which adopted the prorating formula. 

It retained 15 rem as the maximum annual dose and superseded the 13-wk whole-body limit of 3 rem 

with “the provision that not more than one-fourth of the 15-rem maximum permissible yearly dose 

shall be taken in one-fourth of a year” (AEC 1958). 

The quarterly limit of 3 rem or 12 rem/yr replaced the 15 R/yr associated with the weekly limit (NBS 

1958, p. 3, footnote 2). President Eisenhower approved these values in 1960 for Federal agencies. 

AEC Manual Chapter 0524 was reissued in 1963 (AEC 1963) and twice later as U.S. Energy 

Research and Development Administration (ERDA) Manual Chapter 0524 (ERDA 1975, 1977), which 

provided requirements for radiation safety.

In 1957, NBS Handbook 63 specified a dependence of the relative biological effectiveness (RBE) on 

the linear energy transfer (LET) of the charged particles that actually deliver the dose (NBS 1957). 

NBS used this in the Snyder calculations of maximum permissible neutron flux (NBS 1961), and it is

still used in the radiation control regulations for DOE.

At an April 1962 ICRU meeting, the use of the terms RBE and RBE dose in radiation protection was 

criticized and the terms quality factor (QF, now Q) and dose equivalent (DE, now H) were introduced. 

The ICRU recommended the quantity kerma (kinetic energy released per unit ma

In 1971, NCRP Report 39, Basic Radiation Protection Criteria (NCRP 1971a), recommended an 

annual dose limit of 5 rem and the elimination of the quarterly limit. In April 1975, ERDA reissued 

Manual Chapter 0524 (ERDA 1975), which invoked the 5-rem annual dose limits in NCRP Report 39 

and required adding internal and external dose equivalents if both were known. Monitoring was 

required, “where the potential exists for the individual to receive a dose or dose commitment ... in 

excess of 10% of the quarterly standard [3 rem]”. Personnel monitoring equipment for each individual 

was required for external radiation. ERDA (1975, App. p. 10) stated that to achieve optimum 

accuracy, personnel dosimeters should comply with the performance parameters contained in 

American National Standards Institute standards N13.5, N13.7, and N13/42 WG1 Final draft 1974. 

Quality factors from NCRP Report 38 are specified along with the neutron flux density for 100 mrem in 

40 hr as a function of neutron energy (NCRP 1971b). The guidance in NCRP (1971b) for interpolating 

in energy cannot be accomplished with an instrument. The dose equivalent conversion factors 

reported in ICRP 21 (1973) do not present that problem.

ss) in 1962.

In 1971, ICRU defined the quantity dose equivalent index, the maximum value in a 30-cm-diameter 

sphere, for describing ambient radiation fields for radiation protection purposes (ICRU 1971). ICRU 

extended this discussion in Conceptual Basis for the Determination of Dose Equivalent (ICRU 1976), 
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which defined the concept of deep and shallow dose equivalent indexes as those inside a 1-cm depth 

in the sphere and at a depth between 0.07 mm and 10 mm, respectively. A remaining issue was that 

the quantity was measured near the surface of the sphere but applied to the center of the sphere, a 

distance of 14 or 15 cm. In 1980, ICRU identified the deep and shallow dose equivalent indexes as 

restricted indexes (ICRU 1980). In 1985, ICRU Report 39, Determination of Dose Equivalents 

Resulting from External Radiation Sources, introduced the concepts of aligned and expanded fields to 

eliminate issues of field direction and nonuniform fields; it also introduced several dose equivalents: 

Ambient dose equivalent, directional dose equivalent, individual dose equivalent penetrating, and 

individual dose equivalent superficial (ICRU 1985).

ICRP Publications 26 and 30 (1977, 1979) introduced new dose limits and the associated quantity 

effective dose equivalent as averaged over the radiation-sensitive organs of the body. 

In 1981, DOE Order 5480.1A, Chapter XI, “Requirements for Radiation Protection” (DOE 1981a), 

superseded ERDA Manual Chapter 0524 (ERDA 1977). In 1988, DOE Order 5480.11, Radiation 

Protection for Occupational Workers (DOE 1998), superseded DOE Order 5480.1A, Chapter XI. The 

Order adopted much of the language of ICRP Publications 26 and 30 (ICRP 1977, 1979), and the 

monitoring threshold became 100 mrem effective dose equivalent. The order imposed slight changes 

in quality factor value for neutrons in one table, but did not capture those changes in the table of 

permitted neutron flux density.

Because of questions of quality control for dosimetry, the Conference of Radiation Control Program 

Directors encouraged development of a dosimetry accreditation process, which led to the 

development of ANSI N13.11 (HPS 1983b) and the National Voluntary Laboratory Accreditation 

Program (NVLAP). DOE Guidelines for the Calibration of Personnel Dosimeters (Roberson and 

Holbrook 1984) revised the NVLAP processes. Calibration was to the quantities shallow and deep 

dose equivalent (Hs and Hd) and shallow absorbed dose (Ds), which are similar to the individual dose 

equivalent superficial and individual dose equivalent penetrating dose defined in ICRU (1985). These 

quantities were renamed in 1993 to the personal dose equivalent Hp(d) where d is the depth in 

millimeters (0.07 for surface and 10 for deep) from the surface for which the dose is measured (ICRU 

1993). In 1987, DOE Order 5480.15, Department of Energy Laboratory Accreditation Program for 

Personnel Dosimetry, established the DOE Laboratory Accreditation Program (DOELAP) system for 

dosimetry accreditation (DOE 1987). In 1986, Standard for the Performance Testing of Personnel 

Dosimetry Systems specified the measurement of deep and shallow dose equivalents at depths of 10 

mm and 0.07 mm, respectively (DOE 1986a). 

In 1990, the ICRP redefined the concept of dose equivalent to equivalent dose, redefined quality 

factor to radiation weighting factor, and generated new factors (ICRP 1991). These factors, invoked 

in NIOSH (2002), depend on neutron energy at the entrance to the body rather than on secondary 

particle LET where the dose is received. Dose conversion factors for organs and for ambient dose 

equivalent and personal dose equivalent were generated in ICRP Publication 74 (ICRP 1996) and are 

referenced in the external dose implementation guide (NIOSH 2002).

The quantities to be measured and reported by the dosimetry systems have evolved over the last 

50 yr, but the changing definitions had little effect on dosimetry measurements because, for gamma 

radiation, the differences are small. 
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6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1

It was INEEL policy that personnel expected to receive any radiation dose or whose work was 

centered at the site were assigned a radiation monitoring badge. These badges were usually stored 

at the respective operational area entrance security gate for INEEL facilities. Control badges, which 

are used to subtract background radiation, have also been located there. This practice might have led 

to subtracting environmental radiation from site activities, which would have reduced the reported 

doses. Environmental radiation levels have been monitored for most of the life of the INEEL using 

film badges first and thermoluminescent dosimeters (TLDs) later. Table 4-13 in the INL environmental 

dose TBD (ORAUT 2004a) lists the results of this monitoring at facility fence line locations near the 

security gates interpreted as net annual dose (i.e., in addition to natural background), which can be 

added to individual’s dose history or used for unbadged workers on the site.

Site Administrative Practices

Some individuals who might have occasionally visited site facilities but did little work with radiation 

had badges at several different facilities. It is not appropriate to base missed doses on the multiple 

badges issued. During the early years at INEEL, the badge change frequency was not the same for 

all workers. Workers with low probability of exposure were placed on a longer change cycle than 

those with more chance of exposure. Therefore, missed doses should be based on the actual change 

frequency for a person, and the frequency can be determined from the worker’s data package.

The INEEL dosimetry organization developed a set of basic administrative practices in 1951, which 

have changed somewhat as the technologies of ionizing radiation dosimetry and recordkeeping have 

changed. 

DOE provided the dosimetry information for former INEEL workers, which should include a dose 

summary for the employment period and a copy of each weekly, monthly, quarterly, etc. form with the 

work location, so the individual file could be several inches thick in hard copy. Each sheet is redacted 

so only the person of interest’s name and applicable information are visible. This file provides the 

recorded information as to the exchange period for the person for that time period. The following five

figures are a partial example set of redacted dose reporting forms.

From 1951 to 1958, the INEEL dosimetry staff recorded dose daily on a dose card (Figure 6-1), 

rezeroed the pencil ionization chambers worn by workers, and entered the weekly badge result on the 

same card. On this sample, on October 28, November 16, and December 9, 1954, the badges were 

pulled and read in response to high pencil chamber readings. The personnel monitoring badges have 

always been considered more reliable than pencil dosimeters, so after the film badge results became 

available the daily pencil readings were no longer considered doses of record. However, these values 

can be recovered from the earliest forms for a worst-case estimate of dose. In Figure 6-1, the pencil 

readings totaled 820 mR and the badges reported zero for 18 badges.

Figure 6-2 is a report from reading the films in the same period. On three of the five badges, the more 

sensitive open window (O. W.) result was zero, so the shielded film was not read. On the other two, 

the open window and shielded values were at the minimum recorded density of 0.02, which 

corresponded to a 30 mR penetrating dose.

After the pencil ionization chambers were replaced with self-reading pencil dosimeters (which were 

also ionization chambers), the INEEL operational health physics staff would rezero the dosimeters. 

The film reading was automated and results were stored in a computer. The form shown in Figure 6-3 

reports badge reading for May 1959 when badges were exchanged every 2 wk. The column under 
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Figure 6-1. Individual dose reporting form in use until 1958. 

Figure 6-2. Film report form used in 1958. 
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Figure 6-3. Monthly badge reporting form from May 1959.

the P of Personnel is an area designator with the code listed under Location at the bottom of the 

page. Table 6-1 lists the codes for the areas. The next column was unused and dropped somewhat 

later. The next column was for the reason codes listed in Table 6-2. Computer input card codes are 

listed in Table 6-3, “Irregularity Codes,” and in Table 6-4, “Column 20 codes,” respectively.

Table 6-1. Area codes that could be in claimant dose files. 

Area code Description Area code Description

1 AEC Headquarters Bldg 20, 261, 264 TREAT

2 EBR – I 21 LX

3, 034, 035 CFA 22 GCRE

4, 042,045 MTR, TRA 23 OX at OMRE

5, 053, 055 ICPP 24 ARHG

6 NRF 25 No information available

7 TAN (GE) 26, 263,265 EBR -II

8 Services 27 ML-1

9 NX (X is construction) at NRF 28 On-Site Site Survey

10 AX at TAN 29 Off-Site Site Survey

11, 113 CX at CPP 30 ANP Program at SL-1

12 EX at EBR 31 STPF 

13, 133,135 SPERT, PBF 65 ECF

14 OMRE 66 Non-Security

15 SX at SPERT 67 Division of Compliance

16 SL-1 68 STEP

17, 333 MX at MTR 69 LPTF (Phillips & AEC)

18, 814,815 WP, RWMC 71 CADRE (guard force)

19, 772, 775 TAN (Phillips & AEC) 774,776 SMC
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With the advent of computers, the reports are all computer generated with the effect that even though 

many workers were not exposed to neutrons and did not receive neutron dosimeters, zeros are 

entered by the computer in the dose reports. A zero is not an indication that a dosimeter was 

assigned in a computer generated report.

For the early computer records, (~1975-1984) the penetrating and nonpenetrating labels are reversed 

from the data that is printed. The nonpenetrating result is reported first, then the penetrating, and 

then the neutron result.

Table 6-2. Reasons Codes (Column 68-69).

Old Later Years Old Later Years

01 Regular Pull 11 Lost Pencil (or damaged) Visitor HP Request

02 H.P. Request Misc. Pull 12 H.P. Check

03 High Dosimeter Reading Withdrawn 13 Late Pull

04 Recover Lost Badge Termination 14 Withdrawn Badge

05 Ring Reading 15 Termination

06 Wrist Badge Reading H.P. Request 16 Correction

07 Recovered Lost Badge & 

Withdrawn

17 Records Withdrawn

08 Late Pull for Not Available 18 Lost Film Reading

09 Miscellaneous Pull 19 X-Ray Exposure

10 Temporary Film Late Pull resolved by PEQ 20 Experiment Exposure

Table 6-3. Irregularity codes (columns 70-71).

01 Defective Film 12 Dropped in Canal or Reactor

02 Impossible to Read 13

03 Light Leak 14 Not in Area

04 Water Soaked 15

05 16

06 O.W. Shot with X-Ray 17 Old Lot Film

07 Lost in Processing 18 Stuck Film

08 Heat Exposure 19 Not Available

09 Recovered Lost Badge 20 Lost Badge

10 Contaminated Badge 21 No Film

11 Wore Two Badges at one Time

Table 6-4. Column 20 codes.

“X” Master Card 6 Fast Neutron

1 Summary Card 7 Urinalysis

3 Sens. Beta-Gamma 8 Summary Card

4 Insen. Gamma 9 Summary Card

5 Slow Neutron 0 Total Body Results Card

Figure 6-4 is a listing of some doses received during recovery from the SL-1 accident. Workers from

several areas were pulled into the accident recovery process, and it is notable that one result exceeds 

the dose limits and that there are few zeros. Figure 6-5 is a follow-up badge report for one result on 

Figure 6-4. When there was a question about an assigned dose value, a Personnel Exposure 

Questionnaire was normally initiated as shown in Figure 6-6 (shows a hypothetical case).
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 15 of 51

Figure 6-4. Badge pull results from January 1961 for work in recovery from 

the SL-1 accident. 

Figure 6-5. Special badge report associated with a high beta reading listed 

in Figure 6-4. 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 16 of 51

Figure 6-6. Personnel exposure questionnaire partially completed for a 

hypothetical case. 

6.3.2

6.3.2.1 Initial Film Badge

Personnel Monitoring Systems

The badging system in place when operations began at the NRTS was called the Self-Service System 

(Cipperley 1958). This film system, in use from August 1951 to March 1958, used the Oak Ridge 

National Laboratory stainless-steel holder, which was 1.875-in. long, 1.375-in. wide, and 0.25-in. thick. 

Badges were processed weekly. The upper portion of the badge was shielded with 1 mm of cadmium 

and the lower portion was an open window. Sensitive and insensitive DuPont 552 film was used for 

beta-gamma dosimetry for most locations; DuPont 558 film (a combination of types 508 sensitive and 

1290 insensitive films) was used at two reactor areas. 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 17 of 51

Gamma calibration was to a radium source, and beta calibration was to a metallic uranium plate. To 

determine doses, the film densities were read to ±0.02 density unit. A calibration curve was used to 

convert the cadmium-shielded portion to penetrating gamma exposure in roentgen. The open window 

density corresponding to the gamma exposure was subtracted from the measured open window 

density and the remainder was converted to beta dose in rep.

DuPont 552 film has a threshold level of about 30 mR, and DuPont 558 film has a threshold level of 

about 10 mR (Cipperley 1958). The open window responds to beta radiation as well as X-rays and 

low-energy gamma rays. Because of the high atomic number (Z) of film in relation to air or tissue, the 

open window over-responds per unit exposure to low-energy photon radiation, as shown in Figure 

6-7, by about a factor of 30 at 40 keV. Because there was no isolated plutonium at the NRTS, the 

non-penetrating radiation is considered to be beta radiation. Using a cadmium filter with its high Z

severely attenuates the photons that get to the film, so the over-response is reduced to about a factor 

of 2 at 125 keV. The beta particle range is independent of the atomic number Z depending only on 

the density, so the 1-mm cadmium filter (~900 mg/cm2

) simulates a tissue depth of 9 mm for beta 

radiation. 

Wrist badges used the same package attached to a wrist band. A finger ring used a small piece of 

film with a silver or cadmium filter. Pencil ionization chambers were used to monitor daily doses and 

control operational activities. The dosimetry group read these chambers and recorded the results on 

cards. Film badge readings were written on the same cards to indicate the dose of record. In 1958, 

the Victoreen 352 pencil ionization chambers were replaced with self-reading dosimeters that were 

read and rezeroed by the field health physics technicians (Horan 1959, p. 11). Film readings 

remained the dose of record.

6.3.2.2 Multiple-Filter NRTS Film Badge

In March 1958, the security badge and film badge were combined in a film badge containing filters of 

1 mm cadmium, 0.013 mm silver, and 0.5 mm aluminum with thicknesses of 950 mg/cm2

, 

203 mg/cm2

, and 175 mg/cm2, respectively, including the plastic in which they were mounted 

(Cipperley 1968). This NRTS badge was also a security badge, which resulted in an absorber 

thickness of 100 mg/cm2 in the open window that filtered out beta radiation below 360 keV.

With the four absorbers, it was possible to separate beta radiation from photon radiation and to 

somewhat determine photon energy. Figure 6-7 shows photon energy dependence of the darkening 

behind the four filters for a combination of uranium beta and X-ray irradiation provided by NBS 

(Cipperley and Gammill 1959). With DuPont 508 film, mixed exposures of radium gamma and 

uranium beta of 10, 20, and 30 mR or mrep were measurable within ±12 mR with 95% confidence. A 

minimum reporting level of 10 mrem was used for both beta and gamma radiation (Horan 1962).

The Instrument and Development Branch developed an automatic film reader and densitometer 

(Purcell and McGary 1963). An algorithm based on probit-corrected densities was developed to 

determine the contributions of the betas and the high- and low-energy photons separately (Cipperley 

1968, p. 94). The cadmium filter provided the hard gamma component. The uranium beta responses 

under the open window, aluminum, and silver filters were 1, 0.2, and 0.1, respectively. By assuming a 

beta signal and subtracting it, the remaining signal could be attributed to low-energy photons and the 

energy could be estimated. For beta other than uranium, the analysis had greater uncertainty.

Because about 95% of the weekly badge films had doses less than 30 mrem, in 1958 the badging 

interval was increased to biweekly or monthly with the exception of the high-dose areas at which the 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 18 of 51

Figure 6-7. Response of DuPont 508 film with various 

filters to 140 mrem uranium beta and 100 mR of different 

energy photon irradiation. The original badge used the 

open window and cadmium shielded films. The multiple 

filter badge used all three filters plus the open window 

(Cipperley and Gammill 1959).

weekly schedule continued (Horan 1959). The introduction of punch cards increased the efficiency of 

report and record generation. A 12-point calibration curve was generated for radium and for 137Cs 

gamma and uranium beta. Calibration did not use a phantom.

Experience following the SL-1 accident showed a wide variation of beta-to-gamma ratios and 

necessitated controlling both radiations rather than just the gamma. A set of as many as 18 badges 

could, and in many cases was, fastened on a belt around the worker to determine a beta:gamma ratio 

for each particular entry (Cipperley, Henry, and Cusimano 1965). 

6.3.2.3 Original Lithium Fluoride Teflon TLD System

Beginning in November and December 1966, individuals projected to receive doses of less than 

0.5 rem/yr were given a lithium fluoride (LiF) disk TLD badge, which was exchanged quarterly 

(Cusimano and Cipperley 1968). Two 13-mm-diameter Teflon disks, 0.4 mm thick [100 mg (75 

mg/cm2

) impregnated with 28 mg LiF], were mounted in a badge behind an open window and a 1-mm 

cadmium filter (Watkins no date). The disks, manufactured by Teledyne Isotopes, were read with the 

Teledyne Model 7300 TLA reader. LiF was chosen because the average Z is close to that of air and 

tissue, resulting in little energy correction for beta or gamma radiation. The badge could read 30 mR 

on a quarterly basis, so more small doses were reported (Cusimano and Cipperley 1968). The 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 19 of 51

angular dependence of the radium-gamma response (within 10% to 70o) is superior to film because 

the material acts like an ionization chamber. For normal monitoring, only the open window TLD was 

read, and that was considered the penetrating dose unless it read more than 125 mrem, in which case 

the shielded TLD was also read. 

The pilot tests were successful, and the LiF Teflon TLD system was phased into use in 1966, 

particularly for individuals who would receive low doses (i.e., with longer exchange cycles, typically 3 

or 6 months). In July 1968, the monitoring period was increased from 3 to 6 months (Voelz 1969). In

December 1972, annual processing was used for 1,190 low-dose individual TLDs and 960 were 

processed quarterly (Cusimano 1972). Employees on a monthly badge change were moved to this 

system as late as September 1973.

The system had an automatic badge calibrator that did not involve a phantom to provide backscatter 

(Cipperley 1966; Voelz 1970, p. 8). A later discussion introduced the use of a 137Cs source, so these 

earlier calibrations probably used radium.

6.3.2.4 Automatic Thermo Luminescent Analyzer System 

Development began in 1969 on a patented Automatic Thermo Luminescent Analyzer System

(ATLAS). It used LiF in a homogeneous mixture with Teflon and replaced the film in the multielement 

badge using the same filters. This system became operational for monthly badge changes in 

February 1974. In June 1974, questions about this system were formalized (Black 1974; Walker 

1974). A minimum reporting level (MRL) of 30 mR was used for gamma and beta (Walker 1974).

6.3.2.5 Harshaw Two-Chip TLD System

Several unstable characteristics of the ATLAS led to rapid implementation of a two-chip TLD system 

beginning in December 1974 for the Idaho Chemical Processing Plant (ICPP), in February 1975 for 

the prime contractor at TAN, TRA, etc., and in May 1975 for Argonne National Laboratory-West 

(ANL-W). This commercial Harshaw system used two LiF TLDs 240 mg/cm2 thick. In 1976, holes 

were punched in the security badges to restore the open window. One chip was covered by 

540 mg/cm2 of aluminum and the other was under 4 mg/cm2 of Mylar. The aluminum-covered chip 

provided penetrating dose at a nominal tissue depth of 1 cm. The beta dose was calculated from the 

difference between the two chips. Because of the thickness of the Mylar-covered chip, the beta dose 

was accurate only for the beta energy used in calibration. Field calibrations were used to reduce the 

problem with beta energy dependence. The initial thin aluminum filter (density thickness 350 mg/cm2

) 

allowed higher-energy beta radiation to expose the chip used for measuring the penetrating (at 

1,000 mg/cm2

) dose so it was changed to 2 mm in July 1977 (INEL 1978). 

The practice was to read only the open window chip to determine if the nonpenetrating dose was 

above 15 mrem and thus reportable. If the threshold dose was exceeded, both chips would be read 

and the penetrating and nonpenetrating doses computed (Kalbeitzer 1983).

6.3.2.6 Panasonic Four-Chip System

In 1986, with the advent of DOELAP, INEEL changed to the four-element Panasonic 814 AS4 System 

(Gesell, Hall, and Andersen 1992; INEEL 2001). Lithium borate (Li2B4O7) TLD elements with plastic 

and aluminum filtration provide an improved measurement of deep dose equivalent and, with a thinner 

filter, an improved measurement of the shallow dose equivalent. A calcium sulfate (CaSO4) TLD 

provides a strong low-energy photon response. The elements are 15 mg/cm2 thick. Element 1 has 

filtration of 16 mg/cm2

, element 2 has filtration of 58 mg/cm2 plastic, and elements 3 (CaSO4) and 4 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 20 of 51

have filtration of 550 mg/cm2 of plastic and 50 mg/cm2 of aluminum. Although none of the elements is

at a depth of 7 mg/cm2

, the specified depth for the shallow dose equivalent, an acceptable response 

can be obtained by using elements at 16 and 58 mg/cm2

. This system is qualified in DOELAP for

performance categories beta, photon, and mixture. The angular dependence of this system has been 

measured for 137Cs gamma rays and four X-ray energies from 16 to 70 keV (INEEL 2001). For the 

lower energies, attenuation in the absorbers reduces the measured dose. (The attenuation also 

reduces the delivered deep dose, but this effect is not incorporated in the results.) 

The minimum reporting level was 15 mrem beta and gamma from January to July 1986 (Gesell 1986), 

10 mrem gamma and 30 mrem beta from July 1986 to about September 1989, and 15 mrem for 

gamma and 30 mrem for beta until 1993 (Perry, Andersen, and Ruhter 1993), when it returned to 

10 mrem gamma.

6.3.2.7 Nuclear Track Emulsion, Type A for Neutrons

Kodak nuclear track emulsion, type A (NTA) film was used for neutron monitoring when the field 

radiation protection staff requested it. NTA responds to neutrons with energies above 500 to 800 keV, 

for which the proton recoil tracks leave enough energy to expose at least three (four in some 

references) grains of emulsion.

The minimum dose assigned was 14 mrem. Before 1958, if a proton recoil track was counted in 40 

microscope fields, it was read twice more for a total of 120 fields (Cipperley 1958). On one data sheet 

from March 1958 with 10 neutron readings, three persons received 14 mrem and one received 

42 mrem. Two of the four had gamma readings (the MRL was 30 mrem). A blank is recorded for 17 

people on the data sheet, probably because they were not monitored for neutrons or the film was not 

read. Only the two individuals received measurable gamma doses. A person on weekly exchange 

from January to March 1958 received neutron dose equivalents of 14, 28, 42, and 14 mrem and 

gamma doses of 130, 70, 30, 30, 50, 30, and 20 mrem with both neutron and gamma doses received 

twice. These values indicate that the data sheets support the MRL of 14 mrem.

After 1959, if more than three proton recoil tracks were counted in 40 microscope fields, a different 

location on the film was counted by two other technicians, providing three independent results 

(Cipperley 1968). Two tracks or fewer were attributed to background. This resulted in a somewhat 

higher MRL. In November 1959, a data sheet shows a 10- and 20-mrem neutron dose equivalent. In 

January 1962 a data sheet shows a 20-mrem dose. A data sheet from April 1959 shows neutron 

dose equivalents of 20, 20, and 40 mrem. These values suggest an MRL of 20 mrem.

Calibration was with a polonium-beryllium (Po-Be) source (approximately 30 Ci), which resulted in 

5.87 × 10-4 tracks/n (Cusimano 1963). Uncertainties were assigned at the 90% confidence level. 

Cipperley (1968, pp. 102–115) discusses this process.

The field health physics staff was aware of the energy limitations of the NTA badge (Sommers 1967, 

1969) and compensated with neutron-detecting pencil dosimeters and field measurements. A request 

to read NTA film occurred if the hard spectra neutron exposure was likely to exceed 10 mrem. 

Procedures were established using boron trifluoride (BF3) pencils to monitor neutrons in the 

Radioactive Waste Management Complex (RWMC) transuranic waste areas where NTA would not 

respond to low-energy neutrons (Sommers 1975).
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 21 of 51

6.3.2.8 Neutron Albedo Dosimetry

Because of the missed dose from neutrons below the NTA energy threshold of 0.5 to 0.8 MeV, 

particularly at plutonium facilities, and because of the advent of TLD techniques, several development 

efforts in neutron dosimetry occurred in the early 1970s. The results were several designs using the 

albedo technique in which scattered neutrons from the wearer’s body were absorbed by 6

Li in a TLD. 

In the Hankins dosimeter used at the INEEL (Hankins 1973), TLDs with 6

Li to capture thermal 

neutrons are inside a polyethylene case (to lower the neutron energy) inside a cadmium shell (to 

eliminate thermal neutrons from outside) and 7

Li TLDs are used to subtract the gamma dose. 

Because the 6

Li(n,α)3He reaction has a strong energy dependence, the response does not follow the 

flux-to-dose-equivalent conversion, so the neutron signal is divided by a facility neutron correction 

factor (FNCF). A FNCF that converts the TLD gamma equivalent signal to neutron dose equivalent

can be generated from the ratio of the dose equivalent measured with a 9-in.-diameter Eberline 

PNR-4 and the corresponding signal (in millirem but not dose equivalent) with the detector in the 

3-in.-diameter PNR-4 insert. A plot of FNCF versus 9” to 3” ratio is used to determine the FNCF from 

the measured ratio (Hankins 1976). Values of the FNCF as shown in Table 6-5 (Cusimano 1981) 

were measured for different fields at INEEL, were tabulated for assigning the dose equivalent from the 

badge results, and were routinely updated. This correction was applied to generate the reported 

neutron dose. An MRL of 15 mrem was used (Gesell et al. 1996). The angular dependence of this 

system has been measured for moderated 252Cf neutrons (INEEL 2001). 

Table 6-5. 1981 FNCFs.a

Organization FNCF Organization FNCF

DOE-CFA 0.092 EG&G-TRA (Bare PuBe) 0.06

EG&G-CFA 0.092 EG&G-TRA (PuBe in poly) 0.23

ANL-TREAT 1.05 EG&G-LOFT 3.5

ANL-ZPPR 0.94 EG&G-ARA III 2.0

EG&G-TRA (L & S) 1.2 EG&G-RWMC 0.33

EG&G-TRA (SA) 2.7

a. ZPPR = Zero Power Physics Reactor.

The date of the change from NTA to albedo neutron monitoring is somewhat in dispute. Different 

organizations would typically have transitioned to new monitoring systems at different times. The 

present record suggests the switch occurred near the end of 1974 or early 1975 (Ruhter and Perry 

2002; Gesell et al. 1996), but an informal list of “Dosimetry Branch Changes” from 1978 (INEL c.

1978) states, “initial testing of albedo neutron dosimeter and replacement of NTA neutron monitoring 

film with same,” in October 1976. Aoki (1979) says the albedo system replaced the NTA badge in 

1977. Dose reconstructions should make the assumption that this transition occurred on October 1, 

1976.

6.3.3

6.3.3.1 Beta-Gamma Radiation

Calibration

Gamma calibration initially used a radium source. Victoreen R meters standardized by NBS were 

used to measure radiation fields (Horan 1959, p. 132). Uranium metal bars 5 mm thick were used for 

beta calibrations. Cesium-137 was considered for a calibration source in 1959 (AEC 1960, p. 83) and 

was installed in the instrument calibration facility in 1961 (Horan 1962). An automatic badge irradiator 

developed in the 1960s (Cipperley 1966) did not use a phantom to provide backscatter.
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 22 of 51

As reported in 1981, an extrapolation chamber was built for the measurement of beta doses (Gupta 

1981). The chamber window was polycarbonate, the gas was air, and the thick collecting electrode 

was Shonka tissue-equivalent plastic. The chamber was used to calibrate a 2.5 Ci 90Sr/Y source to 

tissue rad. The source with area 2.5 cm2 was constructed by the Amersham Searle Corporation in 

February 1975. This source was used to measure beta correction factors for several instruments 

following the Three Mile Island TMI-2 reactor accident in 1978. TLD badges were calibrated to 

500 mrad tissue using a 1.78-cm-thick phantom 50 cm (300 rad/hr) from the source. 

In January 1983, the natural uranium slab again became the primary calibration source for 

nonpenetrating radiation to better approximate field beta spectra (Gesell 1982a).

Separation of penetrating dose from nonpenetrating dose was an issue in 1957 (Bennett 1957) and 

1976 (Jenson 1976), particularly for ICPP where strong high-energy beta fields were not unusual. 

Use of a phantom in calibration apparently started about 1981 with the NVLAP certification process 

developed for non-DOE dosimetry processors. About this time, calibration developed in terms of 

absorbed dose to tissue rather than exposure. Beginning in January 1981, in response to a draft 

NVLAP (a precursor for DOELAP) standard, dosimeters for calibration were irradiated with 137Cs using 

a phantom backing. To convert from exposure in roentgen to dose equivalent index in rem, a 

conversion factor Cx value of 1.08 was used (DOE 1981b). The current recommended Cx value of 

1.03 for 137Cs (DOE 1986a, Table 2) was used beginning in June 1981 (Gesell 1982b; Kalbeitzer 

1984).

In 1989, the INEEL dosimetry service transferred from the DOE Radiological and Environmental 

Services Laboratory (RESL) to EG&G Idaho, the prime contractor. Calibrations continued to use DOE 

RESL sources and no changes were made to the dosimetry system. The 1991 Tiger Team Review of 

the INEEL site indicated that the INEEL contractor and the Idaho Operations Office using the same 

sources for calibration led to a conflict of interest or an advantage in DOELAP tests. As a result, 

EG&G purchased a Shepherd panoramic irradiator with a 137Cs source for badge irradiations. This 

irradiator did not use a phantom, but was cross-referenced using many TLD irradiations to the DOE 

source using a phantom (Anderson 1995). In addition, the contractor developed and characterized its

own uranium slab for beta irradiations (Bean 1995).

Table 6-6 lists common sources of laboratory bias for personnel beta/photon dosimeter calibration 

based on comparison of the recorded dose with Hp(10). 

6.3.3.2 Neutron Calibration

The initial NTA neutron badges were calibrated using a PoBe neutron source (30 Ci in 1958) (Horan 

1959). In 1982, an AmBe source was used (Cusimano 1982). Alpha particles from the americium or 

polonium interact in the 9

Be(α,n)12C reaction and generate a broad spectrum of neutrons up to about 

11 MeV (mean energy about 5 MeV) as shown in Figure 6-8 (Kluge and Weiss 1982). The yield of 

the AmBe source should be only about 3% larger than that for the PoBe source (Anderson and Hertz 

1971). Kluge and Weiss (1982) calculate conversion factors of 3.51 to 3.76 × 10-8 rem-cm/n

depending on the particular measure of dose equivalent chosen. IAEA (1988) provides a dose 

conversion factor for AmBe of 3.8 × 10-8 rem-cm2

/n for the maximum average dose equivalent. A 

dose equivalent of 1.5 rem required 3.6 × 107

 n/cm2 (Cusimano 1963), corresponding to a dose 

conversion factor of 4.17 × 10-8 rem-cm2

/n, so the recorded dose is about 11% high. Monte Carlo 

calculations for 5-MeV neutrons show a dose equivalent of about 4.2 × 10-8 rem-cm2

/n averaged over 

the 0- to 2-cm shell on a 30-cm-diameter cylindrical phantom (NCRP 1971b). Use of the 50-Ci AmBe 

source continued until 1993.
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 23 of 51

Table 6-6. Laboratory sources of uncertainty for beta/photon dosimeter calibration parameters. 

Parameter Historical description Uncertaintya Comment

In-air 

calibration

In 1981, INEEL began exposing 

calibration dosimeters on 

phantoms (used to simulate 

worker body). Previous 

calibrations do not include 

response from radiation 

backscatter response.

+10% Recorded dose of record too high. 

Backscatter radiation from worker body is 

highly dependent on dosimeter design. 

Radiation 

quantity

Before 1981, INEEL dosimeter 

systems were typically calibrated 

to a photon beam measured as 

exposure.

-5% For higher energy 226Ra and 137Cs gamma 

radiation used to calibrate dosimeters, this 

caused a slight (about 3%) under 

response in recorded dose. 

Tissue depth 

of dose

Historically, INEEL used an 

unspecified depth to estimate the 

deep dose.

±5% The numerical effect of this for photon 

radiation is comparatively low. INEEL 

dosimeter designs had filtration density 

thickness of about 1,000 mg/cm2 that would 

relate closely to the 1-cm depth in tissue.

Angular 

response

INEEL dosimeter system is 

calibrated using anterior-posterior 

(AP) laboratory irradiations.

>300 keV, 

~20%

Recorded dose of record likely too low 

because the dosimeter response is usually 

lower at non-AP angles. Effect is highly 

dependent on dosimeter type, radiation

type, energy, and angle. 

Environmental 

stability

INEEL film dosimeter and TLD 

systems are subject to signal 

fade with time, heat, humidity, 

light, etc.

±10% Recorded dose of record depends strongly 

on dosimetry parameters such as when 

calibration dosimeters were irradiated and 

processed. Mid-cycle calibration minimizes 

effects. 

a. Uncertainty estimate in recorded dose compared to Hp(10) based on judgment.

Figure 6-8. Probability density of neutron spectrum from a 

241Am-Be (α,n) source (Kluge and Weiss 1982). 

In 1993, a 40- by 40- by 15-cm polymethyl methacrylate phantom was placed near the unmoderated 

252Cf source used for instrument calibration, and the system was characterized for TLD calibration 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 24 of 51

(Gesell et al. 1996, Appendix A). This system has since been used for neutron dosimeter quality 

assurance measurements. Calibration factors from the DOELAP manual are used (DOE 1986b).

Table 6-7 describes several common sources of expected laboratory bias for personnel neutron 

dosimeters based on comparison of the recorded dose with Hp(10). 

Table 6-7. Common sources of laboratory bias in the calibration parameters for neutron dosimeters.

a

Parameter Historical description Anticipated laboratory biasb

Source energy 

spectrum

In 1976, INEEL began using dosimeters 

that were calibrated on a phantom to 

simulate a worker’s body. The previous 

calibrations did not include response from 

backscattered radiation. 

NTA film tends to be insensitive to albedo 

neutrons, so probably had minimal effect.

Radiation 

quantity

Neutron dose quantities that were used to 

calibrate INEEL neutron dosimeters have 

varied historically. The first collision dose 

for fast neutrons and a quality factor of 10 

was used for many years. 

As noted above, NTA calibration would result 

in the reported dose being about 11% high. 

The effects of the respective neutron dose 

quantities used to calibrate INEEL 

dosimeters is uncertain and could be 

evaluated in comparison to the Hp(10) dose 

used in DOELAP performance testing. 

Angular response INEEL dosimeters are calibrated using AP 

laboratory irradiation.

Recorded dose of record is likely too low 

because the dosimeter response is lower at 

non-AP angles. The effect is highly 

dependent on neutron energy.

Environmental 

stability

INEEL NTA film and TLD dosimeters are 

subject of signal fade with time, heat, 

humidity, light, etc.

Recorded dose of record is likely too low; 

however, this depends strongly upon when 

the calibration dosimeters are irradiated 

during the dosimeter exchange cycle. 

Mid-cycle calibration minimizes the effects. 

a. Judgment based on INEEL dosimeter response characteristics.

b. Recorded dose compared to Hp(10). 

6.3.4

Radiation fields varied a great deal at the INEEL as discussed in the Site Description TBD (ORAUT 

2005a). At times, very high radiation fields existed, particularly at ICPP, and work was sometimes 

done in those fields under tight controls. 

Workplace Radiation Fields

The epidemiological study of INEEL (Schubauer-Berigan et al. 2005) provides the excellent summary 

of radiation exposure history at the INEEL in Table 6-8. Table 6-8 shows the number of people 

monitored in the last column, the mean dose received (1 mSv=100 mrem), and doses at several 

percentages on the distribution of doses received. For example, in 1952 90% of the 913 monitored 

workers had doses less than 30 mrem and 99.5% of the workers had doses less than 727 mrem. The 

mean dose was 2.2 mrem and the standard deviation of dose was 100.2 mrem. 

The collective dose received by all workers through 1990 is shown in Figure 6-9 taken from Horan 

and Braun (1993). The highest collective dose was 3448 rem in 1965.

6.3.4.1 Gamma Radiation

In response to a Tiger Team finding, radiation fields at the INEEL have been characterized by 

comparing field measurements with a NaI(Tl) gamma spectrometer and TLDs mounted on a phantom 

(Reilly 1998). Figure 6-10 shows the percentage bias for the beta and gamma measurements. Most 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 25 of 51

Table 6-8. Penetrating photon dose (rem) percentiles for monitored workers 

(Schubauer-Berigan et al. 2005). 

Year 0% 25% 50% 75% 90% 97.50% 99.50% 100% MEAN STD 

DEV

NUMBER

1951 0 0 0 0 0 0 0.4 0.4 0.0022 0.0292 188

1952 0 0 0 0 0.03 0.222 0.727 1.275 0.0216 0.1002 913

1953 0 0 0.06 0.25 0.56 1.09 2.097 7.57 0.2002 0.4034 1408

1954 0 0 0 0.21 0.9 2.498 4.077 5.76 0.2838 0.6567 2449

1955 0 0 0.1 0.59 1.6565 3.237 5.162 8.47 0.5157 0.9314 2946

1956 0 0 0 0.2 1.015 2.555 4.267 22.06 0.3142 0.8423 3209

1957 0 0 0 0.085 0.54 1.513 3.045 5.12 0.1706 0.4517 4695

1958 0 0 0.04 0.325 1.26 2.89 4.378 10.51 0.3766 0.8015 5079

1959 0 0 0.04 0.26 0.96 2.347 4.127 21.85 0.3099 0.7402 5344

1960 0 0 0.05 0.255 1.075 2.647 3.734 5.01 0.3268 0.6755 5827

1961 0 0 0.04 0.42 1.704 3.566 4.972 27.26 0.5063 1.2806 5192

1962 0 0 0.03 0.175 1.115 3.225 5.048 9.885 0.3571 0.8942 5339

1963 0 0 0.025 0.18 1.0795 2.954 4.022 5.1 0.318 0.7292 5520

1964 0 0 0.01 0.215 1.2765 3.111 3.98 4.815 0.3538 0.7886 5446

1965 0 0 0 0.43 2.2045 4.39 6.018 9.815 0.579 1.2107 5520

1966 0 0 0.02 0.375 1.5595 3.467 4.466 6.045 0.4383 0.8926 5180

1967 0 0 0 0.17 1.0725 3.084 4.377 4.805 0.3194 0.7702 6304

1968 0 0 0 0.215 1.14 3.104 4.194 5.295 0.3364 0.7781 4922

1969 0 0 0 0.24 1.195 2.73 3.982 4.45 0.3279 0.7151 4758

1970 0 0 0 0.175 1.028 2.624 4.207 4.68 0.2952 0.702 5051

1971 0 0 0 0.17 0.82 1.899 3.252 4.71 0.2357 0.5427 4764

1972 0 0 0 0.161 0.855 2.375 3.835 4.665 0.2606 0.634 4762

1973 0 0 0 0.115 0.6525 2.126 3.909 5.2 0.2185 0.5878 4494

1974 0 0 0 0.1 0.515 1.696 2.993 4.065 0.1734 0.4574 4878

1975 0 0 0.006 0.09 0.405 1.412 2.789 3.945 0.1531 0.4071 5025

1976 0 0 0 0.11 0.506 1.646 2.643 4.145 0.1712 0.4294 5489

1977 0 0 0 0.094 0.485 1.947 3.385 10.77 0.1869 0.526 5677

1978 0 0 0.009 0.087 0.394 1.535 2.872 4.386 0.1563 0.4295 6551

1979 0 0 0 0.074 0.369 1.46 2.732 4.18 0.1419 0.4064 6863

1980 0 0 0 0.047 0.277 1.062 2.112 16.93 0.1073 0.3579 7380

1981 0 0 0 0.046 0.252 0.824 1.674 3.289 0.0876 0.2358 6722

1982 0 0 0 0.038 0.195 0.638 1.5 2.904 0.0715 0.204 6556

1983 0 0 0 0.035 0.187 0.516 0.976 1.577 0.0582 0.1473 6610

1984 0 0 0 0.034 0.179 0.577 1.198 2.285 0.0619 0.1766 7476

1985 0 0 0 0.039 0.219 0.872 1.24 2.415 0.082 0.226 7917

1986 0 0 0 0.024 0.192 0.773 1.749 9.338 0.077 0.2623 8568

1987 0 0 0 0.025 0.155 0.733 1.417 3.158 0.0659 0.2098 8575

1988 0 0 0 0.022 0.145 0.537 1.086 3.086 0.0545 0.1641 8667

1989 0 0 0 0.016 0.101 0.559 1.315 7.811 0.0516 0.2169 8848

1990 0 0 0 0.012 0.0954 0.566 1.266 2.728 0.049 0.1852 10165

1991 0 0 0 0 0.037 0.249 0.694 4.577 0.0232 0.1094 10742

1992 0 0 0 0 0.037 0.169 0.376 1.276 0.0157 0.0563 9571

1993 0 0 0 0 0.065 0.342 1.06 1.535 0.0311 0.1233 9048

1994 0 0 0 0.01 0.0786 0.335 0.729 1.394 0.0322 0.1121 8473

1995 0 0 0 0.011 0.1031 0.449 1.153 1.844 0.0428 0.1533 7818

1996 0 0 0 0.016 0.108 0.421 0.79 1.368 0.0393 0.1215 6459

1997 0 0 0 0.01 0.081 0.259 0.574 1.108 0.0261 0.082 6280

1998 0 0 0 0.002 0.062 0.208 0.462 0.844 0.0205 0.065 5875

results lie within +27% to -43%. The high gamma bias results are for locations at RWMC looking at

skyshine (back-scattered, low-energy photons) from low-level waste in the Subsurface Disposal Area. 

The doses measured with NaI(Tl) were low (6 and 11 mrem) and the threshold energy on the NaI(Tl) 

detector was about 100 keV, so some low-energy photons were likely to have been missed. The 

radiation fields at INEEL, with a few exceptions, are generated primarily by mixed fission and 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 26 of 51

activation products. Therefore, most of the photon dose is from photons with energy greater than 

250 keV. The INEEL dosimeters are judged to measure these fields well.

Figure 6-9. Collective dose for INEL personnel from 1951-1990 

(Horan & Braun 1993). 

 

0

1

2

3

4

5

6

7

-85 to -75%-75 to -65%-65 to -55%-55 to -45%

-45 to-35%

-35 to -25%-25 to -15%

-15 to -5%

-5 to 5%

5 to 15%

1'5 to 25%

25 to 35%35 to 45%45 to 55%55 to 65%65 to 75%75 to 85%85 to 95%

95% to 105%

105% to 115%115% to 125%125% to 135%

Relative Bias [(Dosimeter-Spectrometer)/Spectrometer]

Number

Gamma

Beta

Figure 6-10. Gamma and beta radiation field characterization.

The few exceptions are usually characterized by low dose rates. Much of the waste at the RWMC is 

transuranic waste from the Rocky Flats Plant. This contains predominantly plutonium and americium 

with 17- and 59-keV photons, which are highly adsorbed by the waste and the shipping container. 

The dominant fields at the RWMC come from mixed fission and activation products dominated by 

137Cs and 60Co.
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 27 of 51

Analytical X-ray generators operating below 100 keV are used in several laboratories. These are 

easily shielded so the fields are usually low.

There are a few 250-keV x-ray generators used for radiography or radiography development studies. 

Wall shielding is generally adequate and any transmitted photons have energy near the operating 

voltage because of the hardening caused by the shielding. The radiography facility at TRA has no 

roof shielding, so dose rates nearby of less than 10 mrem/hr are possible, usually for short periods. 

The scattered photons have lower energy than the primary beam. The people likely to be exposed 

are radiographers who receive fairly high exposures, mostly from the ambient radiation fields at their 

worksite and some from 300-keV 192Ir or 1.25-MeV 60Co.

Essentially all INEEL radiological work areas involved beta and photon radiation covering a wide 

range of energies. These fields can be generally classified according to the IREP codes in Table 6-9. 

Table 6-9. Selection of IREP beta and photon energies for INEEL facilities.

a

Process/

buildings Description

Operations Radiation 

type

Energy 

Begin End group (keV) Percentage

Reactors Highly dispersed fields of higher energy photon radiation fields 

from fission process, activation and fission product nuclides. 

Potential for significant airborne nuclides, and there might be 

significant higher energy beta radiation

Beta >15 100

Photon 30–250 25

> 250 75

PBF, TRA, ARA, TAN, EBR, ANL-W, SPERT

1952 2003

Processing 

plants 

Highly dispersed fields of higher energy photon radiation fields 

from activation and fission product nuclides dominant to most 

exposure profiles. Potential for higher energy beta radiation 

during sampling and maintenance work resulting from fission 

products. 

Beta > 15 100

Photon 30–250 25

> 250 75

ICPP 1952 2003

Calibrations Calibration of instruments and dosimeters Beta > 15 100

CF 633, 636 1952 2003 Photon 30–250 25

> 250 75

Waste 

handling

Radiation characteristics are highly dependent on source of waste, 

but typically fission product nuclides (Sr/Y-90, Cs-137) are 

dominant. For transuranic waste from Rocky Flats contains 

Am-241 with 59-keV photon.

Beta > 15 100

Photon 30–250

> 250

25

75

RWMC, WERF 1953 2003

Uranium 

handling

Produced special armor from depleted uranium. Primarily beta 

radiation from U-238 daughters. Some gamma from contaminants 

and Cs-137 sources used in process.

Beta > 15 100

Photon 30–250 90

> 250 10

SMC 1985 2003

a. PBF = Power Burst Facility; SMC = Special Manufacturing Capability.

6.3.4.2 Beta Radiation

Beta radiation fields are usually associated with activation or fission products outside of a container 

such as in a spill or only lightly shielded or in hot cells. High beta fields were not unusual at the Idaho 

Nuclear Technology and Engineering Center (INTEC) where large quantities of fission products exist. 

Pure high-energy beta fields in some locations, particularly in the exhaust stream, have caused 

dosimetry problems because the badge shielding or instrument packages did not provide a full 10-mm 

tissue equivalent coverage and the beta fields would therefore be measured as gamma fields.

The high bias beta results in Figure 6-10 from comparison of TLDs to a phoswich beta spectrometer 

are for sources at contact or at 1 cm, which results in hard to reproduce geometry. The low bias beta 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 28 of 51

results are for large area sources for which even the spectrometry results have large variations. The 

beta occupational radiation fields (only three) have a bias better than 15%.

Beta field dosimetry became fairly accurate with the definition of DOELAP requirements in the early 

1980s. Before then, beta monitoring systems had various flaws, primarily in a detector too thick to 

give a good surface result or one that was covered with extra material. Calibration was to high-energy 

betas from either uranium or strontium. The dose from low-energy betas is missed altogether if the 

beta energy is not sufficient to penetrate the detector cover and is underreported if the beta energy is 

not sufficient to penetrate the entire detector. The mean beta energy for the spectrum from a 

particular nuclide is about one-third of the maximum beta energy for that nuclide. 

Based on the range vs. energy curve for beta particles and the beta energy distribution of beta 

emitters (BRH 1970, pp. 90, 91,123), the fraction of beta radionuclides with ranges greater than the 

abscissa is plotted on the ordinate in Figure 6-11. Beta-emitting nuclides varied from location to 

location and time to time at the INEEL, so a correction factor common for all facilities was estimated. 

 

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 50 100 150 200 250 300

Range (mg/cm^2)

Fraction of Nuclides

Max energy

Max 45 Incid

Mean Energy

Mean 45 Incid

Figure 6-11. Distribution of beta ranges.

This analysis used the entire mixture of radionuclides to obviate whether the choice is correct and to 

reflect the wide variety of radionuclides used at the INEEL. To reflect that the beta spectrum is not 

monoenergetic because of the energy carried off by the neutrino, a curve is presented for the mean 

energy or one-third of the maximum energy. To reflect that some beta particles enter the detector at 

an angle, a curve is provided for 45o incidence at the maximum energy and the mean energy. These 

curves of the fraction of nuclides with a larger range essentially show the depth dependence of beta 

dose, because the energy loss of electrons does not have much energy dependence. These curves 

also demonstrate why early dosimeters with thicker more sensitive elements failed to report the beta 

dose correctly at a depth of 7 mg/cm2

, a depth that was chosen in the early 1980s. These curves 

demonstrate why the beta dose assigned for skin is inappropriate to use for the breast and testes 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 29 of 51

where much of the organ is at a depth greater than 1 mm or 100 mg/cm2, and for most of us at depths 

greater than 1 cm.

To calculate the fraction of dose missed by a dosimeter, the dose reconstructor need only to average 

the appropriate curve of this nature over the depth of the active detector and compare it to the value 

at a depth of 7 mg/cm2

. The appropriate curve should be the curve of the mean range for the beta 

spectrum and the angular distribution of the radiation exposure. To estimate this, this analysis added

the mean energy curve for perpendicular incidence and 1.4 (relative path length) times that for 45o 

incidence for the mean energy and to add in one-half that value for 45o incidence for the maximum 

energy. The curves are the result of a polynomial trend line to the data, so averaging the fraction of 

radionuclides is relatively easy.

Table 6-10 provides the cover and detector thicknesses for the beta badges used at INEEL. 

Thicknesses with a “~” are estimated. The fraction of measured beta dose shown in Table 6-10 is the 

average as described above. To determine the corrected beta dose, the measured and missed 

non-penetrating result from the dosimetry system should be multiplied by the values in the last column 

of Table 6-10. The reported dose will likely be somewhat higher than this because the calibration 

probably did not consider such a correction and reported the dose for the calibration exposure. For 

the Panasonic system, such a correction has already been made (Gesell 1986), so the recommended 

correction is 1.

Table 6-10. Beta dosimeter thicknesses and associated underreporting. 

Dosimeter 

system Period

Covers

(mg/cm2)

Detector thickness

(mg/cm2)

Beta Correction 

Factor

Two filter film 1951-1958 ~50 ~50 2.0

Multifilter film 1958-1974 100 ~50 2.8

Low dose TLD 1969-1974 100 75 3.0

ATLAS 1974-1975 100 ~100 3.3

Harshaw TLD 1975-1976 104 240 4.8

Harshaw TLD 1976-1985 4 240 2.4

Panasonic TLD 1986-2006 16 15 1.0

6.3.4.3 Neutron Radiation

Most INEEL workers have not been exposed to neutrons and so have not been badged to measure 

neutrons. Neutron fields at the INEEL have been specific to a few facilities. The high dose-locations 

where most of the gamma and beta dose has been received, such as the ICPP and SL-1, have not 

had associated neutron dose. Table 6-11 lists the facilities and time periods where neutron exposure 

is considered likely.

In 1969, 150 workers were involved in radiation work that required their NTA neutron dosimeters to be 

evaluated. At the time, there were 2,900 film-badged employees and more than 3,000 TLD-monitored 

personnel (Vallario, Hankins, and Unruh 1969).

For calendar year 1979, 5 people received neutron doses between 0.5 and 1 rem and 79 received 

measurable neutron doses below 0.5 rem (Jones 1980). 

Individuals who have the potential to receive neutron dose currently wear albedo badges, and 

experience has shown that most do not receive significant doses. In the first 9 months of 1995, only 

1,461 neutron dosimeters were issued (both monthly and quarterly badges) in comparison with about 

50,000 beta/gamma badges. Only 54 badges had reportable doses (≥15 mrem) as shown in 

Figure 6-12 (Gesell et al. 1996). Only six were above 35 mrem. The Hankins albedo dosimeter 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 30 of 51

badges in use since 1975 are sensitive to all neutron fields. An FNCF determined from the 9- to 3-in. 

ratio in the worker location is used to adjust the measurement result to dose equivalent.

Table 6-11. Facilities and time periods for neutron exposure. 

Location Time period Comments

Programmatic

TREAT 1958-1994

ZPPR 1969-1992 Between reactor halves

MTR 1952-1970 Research Floor

TAN Warm Shop 1986-1988 See EPRI reports

TAN 1986-2006 Spent fuel storage pad

ANL Neutron 

Radiography Facility 

1977-2006

TRA Hot Cell Cave Cf-252 on filters

RWMC ILTSF Cargo Container

RWMC WM 632 TRU Waste Drum

14 MeV

RWMC SWEPP 1990-2004 14 MeV for waste characterization

TAN Warm Shop 1991-1994 Refurbish 14 MeV for waste characterization

TRA 635 1990-2006 PINS Cf-252 6 & 14 MeV

Sealed sources

CPP-1649 1985-2006 PuBe Calibration Facility

CF-633 ~1970-2002 Cf-252 Calibration Facility

CF-636 1952-1994 AmBe Calibration Facility

IRC AmBe Cf-252

 

Figure 6-12. Distribution of reportable neutron dose for the 

first 9 months of 1995. Of 1,461 dosimeters, 1,407 were 

below the 15 mrem reporting level. 

In 1997, several workplace neutron fields were measured with TLDs mounted on a phantom and at 

nearly the same time a ROSPEC neutron spectrometer (Reilly 1998). The relative biases 

[(Dosimeter-Spectrometer)/Spectrometer] for the neutron fields are shown in Figure 6-13. These 

Total number of dosimeters with reportable dose = 54 Total number of dosimeters with reportable dose = 54
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 31 of 51

 

0

1

2

3

4

-55 to -45%

-45 to-35%

-35 to -25%-25 to -15%

-15 to -5%

-5 to 5%

5 to 15%

15 to 25%25 to 35%35 to 45%45 to 55%55 to 65%65 to 75%75 to 85%85 to 95%

Relative Bias [(Dosimeter-Spectrometer)/Spectrometer]

Number

Neutron

Figure 6-13. Neutron radiation field characterization. 

results show a greater dispersion than the gamma results. The two lowest values (-52% and -51%) 

are for TLD measurements on opposite sides of a phantom where the field is from 252Cf on an 

overhead filter bank. The phantom attenuates the radiation from each side so the TLDs only see half 

the radiation field. The next lowest value (-38%) is for the 252Cf instrument calibration source at a 

distance of 3.5 m where the operator stands. The two highest values (0.94 and 0.71) are for a waste 

drum that was reanalyzed and a new 9”:3” ratio determined because of the unsatisfactory initial result. 

The report suggests that other waste barrels might have had neutron sources causing interference. 

The remaining bias values lie between -0.16 and 0.44.

Figure 6-14 provides spectra for the 14-MeV neutron generator as seen through 10 cm of

polyethylene shielding typical of the INEEL facilities, 252Cf (fission) neutron source (Ing and Makra 

 

0.0001

0.001

0.01

0.1

1

0.000001 0.00001 0.0001 0.001 0.01 0.1 1 10 100

Neutron Energy (MeV)

Number neutrons

Fission

14 MeV

AmBe

Figure 6-14. Neutron spectra simulating INEEL facilities. 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 32 of 51

1978) and the AmBe neutron source (Kluge and Weiss 1982). Sources of neutron exposure include 

neutron sources at the instrument calibration laboratories and 14-MeV neutron generators used to 

characterize waste. For these spectra, the NTA works reasonably well. Use of small 252Cf sources for 

research began after albedo badge use began. 

Most of the reactors built at the INEEL had no beam ports. Thus the neutrons were generally well 

contained away from the workplace. The reactor core environment is characterized not only by high 

neutron levels, but also by very high gamma levels. The gamma shielding is often water and concrete 

which are also very good neutron shields. The neutron fields in the energy spectrum for reactors (and 

lower energy) will be attenuated much more quickly in concrete or water than will the gamma fields. 

This is not true for lead or iron, but these are usually not used as gamma shields where neutrons also 

exist. Thus neutron fields are generally not a problem at an enclosed reactor.

6.3.4.4 Materials Test Reactor Neutron Radiation

The exception to the above discussion is the Materials Test Reactor (MTR), which operated from 

1952 to 1970 and had beam ports and neutron beams extending onto a research floor. ZPPR and 

TREAT, both at ANL-W, are also in this category. Some neutron surveys of the MTR experimental 

floor have been recovered (Sommers 1959, 1962; Hankins 1961), but these individually do not 

provide all components of the radiation field. Hankins (1961) used 2-, 3-, and 8-in. polyethylene 

Bonner balls in a cadmium shield to characterize the intermediate and fast neutrons at 21 locations 

around the MTR floor and measured the thermal neutron component at six other locations. The 

Hankins data have been reanalyzed (ORAUT 2004b) using more recent Bonner response curves 

(Hertel and Davidson 1985). Figure 6-15 shows the resultant neutron spectra for locations 3 and 23, 

which have higher doses and nearly the maximum low-energy intermediate and fission components, 

respectively. Figure 6-16 shows the correlations of the thermal and intermediate neutron dose 

equivalents to the fast neutron dose equivalent for the ORAUT (2004) reanalysis of the Hankins data. 

10

100

1000

10000

100000

1.E-01 1.E+01 1.E+03 1.E+05 1.E+07

Energy (eV)

Number Neutrons

Position 23

Position 3

Figure 6-15. Sample MTR spectra from 

Hankins Bonner measurements. 

 

y = 4.633x

R2

 = 0.501

y = 0.148x

R2

 = 0.914

y = 0.049x

R2

 = 0.701

y = 0.115x

R2

 = 0.852

0.01

0.1

1

10

100

0.01 0.1 1 10

Fast Neutron Dose Equivalent Rate (mrem/hr)

Dose Equivalent Rate (mrem/hr)

InterHankins

Inter .5 eV-10 keV

Inter 10-200 keV

ThermalNeut

Figure 6-16. MTR neutron field components.
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 33 of 51

The trend line for the reanalyzed intermediate-energy neutron dose equivalent has an R2 value of 0.86 

and 0.92 in comparison with an R2 value of 0.5 for the original analysis, which demonstrates a better 

fit to the data. The average ratio of thermal to fast neutron is 0.071 ±0.025, for the low-energy 

intermediate to the fast is 0.177 ±0.057, and for the higher energy intermediate to the fast is 

0.149 ±0.046 where fast neutrons are taken as those above 0.2 MeV. 

The MTR personnel who were likely to receive neutron dose were assigned NTA film in their 

dosimetry packets, but it would have missed the dose below 0.5 to 0.8 MeV. For the MTR spectra, 

the fraction of neutron dose equivalent above 0.8 MeV has an average value of 0.52 ±0.08 and varies 

from 35% to 66% depending on the location. The dosimetry record location code for the TRA was 4

(later 40 to 45). To correct for missed dose on the MTR experiment floor, the NTA results from MTR 

should be multiplied by 2 ±0.2 (1/0.52, 0.08/0.52) for a Monte Carlo dose reconstruction. 

Sommers (1962) reported thermal and fast neutron dose equivalent rates and gamma dose rates 

around the MTR beam lines. The thermal measurements near beams are believed not to be 

representative of the general workplace. Figure 6-17 shows the correlation of fast neutron dose 

equivalent to the gamma dose for these measurements. 

 

 

0.1

1

10

100

0.1 1 10 100 1000

Gamma DE Rate(mrem/hr)

Fast Neutron DE Rate (mrem/hr)

Fast Neutron

Fast Insig

Fit 0.419 *Gamma

Figure 6-17. Correlation of fast neutron dose 

equivalent to gamma dose at MTR. 

The fast neutron component was insignificant for several of the measurements. These values are 

shown with the triangles at one-half of the smallest measured value. Using the Shapiro-Wilks 

Normality Test (Gilbert 1987) and including the insignificant fast neutron values at one-half of the 

minimum reported value suggests that the normal distribution is a slightly better description of the 

data than a lognormal distribution. The fast neutron dose equivalent is 0.42 ±0.35 of the gamma dose 

rate for this data set. Combining these results as shown in equation 6-1 yields a total neutron dose 

equivalent of 0.58 ±0.48 of the gamma dose equivalent on the MTR experimental floor: 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 34 of 51

.0( 42 .0 35)( .01 071 .0 177 .0 149 .0) 58 .0 48

1( )

±=+++±=

= + ++

Fast

HiInter

Fast

LoInter

Fast

Thermal

Gamma

FastN

GammaDose

TotalNeutronDose

(6-1)

The variations within the different components of neutron dose rate are so small in comparison with

the variation between the fast neutron and gamma dose equivalent rate that they are unimportant. 

Many of the people wearing NTA film would receive gamma dose at locations other than on the MTR 

experimental floor while the reactor was operating. For example, health physics technicians would 

often have been covering jobs with only beta/gamma fields. A craftsperson might have serviced 

pumps carrying radioactive water and not received any neutron dose. Therefore, simply multiplying 

the gamma dose by 1.6 (i.e. average = 1+0.58) or 2.1 (i.e. 84% confidence = 1+0.58 +0.48), although 

favorable to claimants, is probably inappropriate.

6.3.4.5 Test Area North Fuel Storage Casks

As noted in the INEEL Site Description TBD (ORAUT 2005a, Section 2.2.1.3), fuel storage casks are 

on a storage pad at TAN. The dose rates are 25 to 30 mrem/hr gamma and 40 mrem/hr neutron. 

The metal cask attenuates the gamma radiation, but does not appreciably affect the neutron field. 

The loaded casks were temperature-tested in the TAN Warm Shop in 1985 and 1986. The shop has 

offices nearby at one end on the second floor. Neutron radiation levels were discovered in the offices, 

and the people (about six) in these offices were not wearing albedo neutron badges but were wearing 

beta/gamma dosimetry. Each of the three casks was in the area about 2 wk while temperature 

measurements were made before they were moved outside onto the storage pad. The casks contain 

irradiated oxide fuel in which 18O(α,n) reaction generates neutrons in the range of a few MeV. This 

radiation was attenuated by distance and the building concrete between the cask and the offices. A 

TLD area-monitoring albedo system identified the radiation field. Based on Electric Power Research 

Institute (EPRI) documents (EPRI 1986, 1987; PNL, VPC, and EG&G 1987) the temperatures were 

measured from September 11 to 23, 1985, January 14 to February 6, 1986, and June 2 to 27, 1986. 

The office most affected was for an operations person. Safety- and radiological- engineers were also 

affected. The estimated dose equivalent for full-time occupancy is less than 50 mrem for each of the 

three cask evolutions. 

6.3.4.6 Typical Workplace Neutron Dosimeter Hp(10) Performance

Table 6-12 summarizes typical neutron personnel dosimeter parameters important to Hp(10)

performance in the workplace. The most important parameter related to Hp(10) performance of the 

neutron dosimeters is the difference between calibration and workplace neutron energy spectra. 

6.4 ADJUSTMENTS TO RECORDED DOSE 

6.4.1

All dose equivalents measured at the INEEL and reported in this document used the quality factors 

based on the LET of the ionizing secondary particles established in the 1950s and used since by U.S. 

regulatory agencies. In 1990 the ICRP developed new dose concepts that have been used by 

Neutron Weighting Factor
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 35 of 51

Table 6-12. Typical workplace neutron dosimeter Hp(10) performance.a

Parameter Description Potential workplace biasb

Workplace neutron 

energy spectra

NTA dosimeter response decreases 

and TLD response increases with 

decreasing neutron energy 

Depends upon workplace neutron spectra. 

NTA recorded dose of record likely too 

low because of high 500-keV threshold for 

detection of neutrons. 

Exposure geometry NTA dosimeter response increases 

with increasing exposure angle and 

TLD response decreases with 

increasing exposure angle.

NTA recorded dose likely too high

because dosimeter response is higher at 

angles other than AP. TLD recorded dose 

is lower at angles other than AP. Effect is 

highly dependent on neutron energy. 

Missed dose Doses less than MRL recorded as zero 

dose.

Recorded dose of record is likely too low. 

The impact of missed dose is greatest in 

earlier years because of the higher MRLs 

and shorter exchange cycle of the neutron 

dosimeters. 

Environmental effects Workplace environment (heat, 

humidity, etc.) fades the dosimeter 

signal.

Recorded dose of record is likely too low.

a. Judgment based on INEEL dosimeter response characteristics.

b. Recorded dose compared to Hp(10). 

NIOSH. The quality factor Q as a function of LET was replaced with a radiation weighting factor wR, 

which is a function of the neutron energy (ICRP 1991, Table 1). 

The INEEL reported data requires correction to change from dose equivalent (pre-ICRP 60) to the

newer dose quantity (ICRP 1991; NIOSH 2002). ICRP 74 (1996) tabulates the ambient dose 

equivalent (dose equivalent at 10 mm depth in a 30-cm-diameter sphere) for neutrons. The ratios of 

organ to ambient dose equivalents are tabulated in the NIOSH External Dose Reconstruction 

Implementation Guideline (NIOSH 2002), so this quantity is used for the conversion. Ambient dose 

equivalent is an ICRU quantity so it uses a revised Q(L) rather than a wR, so the correction factors are 

not as large as in other TBDs.

The dose equivalent for a spectrum of particle energies is the result of an integral of the fluence 

spectrum Φ(E) times a dose equivalent conversion factor DECF(E), which also depends on energy 

over the range of energies considered:

H DECF dEEE

E

E

∫ =

2

1

φ )()( (6-2)

Error! Bookmark not defined.

For ambient dose equivalent, the same equation applies except that a tabulation of the ambient dose 

equivalent dose conversion factor is used (ICRP 1996). The correction factor for an energy interval is 

These factors are incorporated into statements of dose equivalent 

values and calibrations following generally accepted principles. The conventional dose conversion 

factors are most clearly and correctly stated in ICRP Publication 21 (ICRP 1973). NCRP Report 38 

tabulates a neutron flux density associated with the annual dose limit that is proportional to the 

reciprocal of the dose conversion factor (NCRP 1971b). The primary geometry is conventionally

considered from one direction with the maximum dose in the body tabulated. More recent references 

(ICRU 1985; ICRP 1988, 1996) consider the dose to individual organs for different irradiation 

geometries, so the more recent tabulations give results lower by factors up to about 10 due to 

attenuation in the human body. Dosimeters are designed to respond to radiation entering the body on 

the side where they are located, and work best for an AP irradiation geometry with the dosimeter on 

the front of the body.
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 36 of 51

then the ratio of the two integrals. Because IREP uses different radiation effectiveness factors for 

different radiation types and energies, it is appropriate to use the IREP energy intervals for calculating 

the correction factors. 

Table 6-13 summarizes the locations at the INEEL where neutron dose is credible. Table 6-13 lists 

the calculated fractions of dose equivalent in the IREP energy groups and the conversion factors from 

dose equivalent to equivalent dose for INEEL spectra. The ratios of average radiation weighting 

factor to average quality factor for the IREP energy groups have some variation, particularly for the 

10- to 100-keV group where the energy dependence of the fluence is radically different for the fission 

and 14-MeV source than for the reactor spectrum. The lower part of the table lists the recommended 

default values for the dose equivalent fractions and quality factor corrections.

Table 6-13. Calculated and recommended dose equivalent fractions and quality factor corrections. 

IREP energy interval < 10 keV 10 keV-100 keV 100 keV-2 MeV 2 MeV-20 MeV

Spectrum calculated values

Dose equivalent fractions

Bare fission 4.4E-05 0.20 0.80

AmBe 0.15 0.85

14 MeV 10 cm poly 2.4E-08 3.1E-06 1.5E-03 1.00

MTR exp floor ave 0.18 0.06 0.49 0.28

MTR exp floor max 0.24 0.08 0.52 0.35

MTR exp floor min 0.13 0.03 0.46 0.19

ICRP 74 H*10 /NCRP 38 H

Bare fission 1.46 1.32 1.09

AmBe 1.41 1.05

14 MeV 10 cm poly 0.69 1.47 1.36 0.93

MTR exp floor ave 0.86 1.08 1.33 1.12

MTR exp floor max 0.80 1.08 1.37 1.12

MTR exp floor min 0.92 1.08 1.30 1.12

Recommended defaults

Dose equivalent fractions

14 MeV 10 cm poly 0.05 0.95

Source calibrations 0.20 0.80

MTR exp floor 0.2 0.05 0.50 0.25

H*(10)/H 1 1.1 1.4 1.1

6.5 MISSED DOSE

6.5.1

Workers have reported that, on some occasions, they did not wear dosimeters while working in 

radiation areas (Wages et al. 1998; Worker Outreach meeting, April 28, 2004). The latest revision of 

Use of Coworker Dosimetry Data for External Dose Assignment (ORAUT 2005b) provides guidance to 

assist dose reconstructors in evaluating the support for a claimant’s allegation that he or she did not 

wear a dosimetry badge at all times. 

Dosimeter Not Worn

6.5.2

Missed photon dose for INEEL workers would occur where a zero dose was recorded for the 

dosimeter systems for any response less than the site dose recording threshold (the MRL). The 

missed dose for dosimeter results less than the MRL is particularly important for earlier years when 

MRLs were higher and dosimeter exchange was more frequent. The missed dose is calculated as

Photon Missed Dose
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 37 of 51

described in NIOSH (2002) using MRL/2 multiplied by the number of zero dose results. Table 6-14

lists the potential missed photon dose by year, dosimeter type, and badge exchange frequency. The 

MRLs shown are based on Cipperley (1958, 1968) and Cusimano (1963) for film; Kalbeitzer (1983), 

Gesell (1986), Gesell, Hall, and Anderson 1992), and Perry, Anderson, and Ruhter (1993) for TLDs; 

and Ruhter and Perry (2002) for film and TLD. The exchange frequency must be determined from the 

individual worker’s dose submittal package for each year because it was shorter for highly exposed 

individuals and longer for those with lower doses.

Table 6-14. Beta/photon dosimeter period of use, type, MRL, exchange frequency, and potential 

annual missed dose. 

Period of usea Dosimeter

Exchange 

frequency

MRLb (mrem)

Annual missed dose 

(mrem)c

Photon Beta Photon Beta

August 1951–

March 1958

INEEL Initial Film 552 DuPont

Film

Weekly (n=52) 30 30 780 780

Monthly (n=12) 180 180

Reactor Areas DuPont 558 Film Weekly (n=52) 10 30 260 780

March 1958 –

December 

1966

INEEL Multi-Element DuPont

508 Film

Weekly (n=52) 10 30 260 780

Biweekly (n=26) 130 390

Monthly (n=12) 60 180

December 

1966–February 

1974

INEEL Multi-Element DuPont

508 Film

Weekly (n=52) 10 30 260 780

Biweekly (n=26) 130 390

Monthly (n=12) 60 180

INEEL LiF TLD Quarterly (n=4) 15 15 30 30

Semi-ann (n=2) 15 15

Annual (n=1) 7.5 7.5

February 1974–

May 1975d

INEEL Atlas TLD LiF in Teflon Monthly (n=12) 30 30 180 180

Quarterly (n=4) 60 60

Semi-ann(n=2) 30 30

Annual (n=1) 15 15

December 

1974-December 

1985d 

INEEL Harshaw Two-chip TLD Monthly (n=12) 15 15 90 90

Quarterly (n=4) 30 30

Annual (n=1) 7.5 7.5

January 1986 -

2006

INEEL Panasonic Four-chip 

TLD 

Monthly (n=12) 15e 15f 90 90

Quarterly (n=4) 30 30

Monthly (n=12) 10e 30f 60 180

Quarterly (n=4) 20 60

a. For many years, INEEL workers had a dosimeter assigned to each operating area where they worked, or were issued 

visitor dosimetry. All Area dosimetry was issued beginning in January 2000.

b. MRLs are based on Cipperley (1958), Cipperley (1968), Cusimano (1963), Kalbeitzer (1983), Gesell (1986), Gesell, Hall 

and Anderson (1992), Perry, Anderson, and Ruhter (1993), and Ruhter and Perry (2002).

c. Maximum annual missed dose calculated using N × MRL/2 from NIOSH (2002).

d. ICPP began using the Harshaw in December 1974, the prime contractor in February 1975, and ANL-W in May 1975.

e. The MRL was 15 mrem from January 1, 1986, to July 7 1986, 10 mrem from July 7, 1986 to about September 1989, and 

15 mrem until 1993 when it returned to 10 mrem.

f. The MRL was 15 mrem from January 1, 1986, to July 7 1986, and 30 mrem after that.

6.5.3

Beta dose is important for certain cancers. Beta electrons are always above 15 keV if they are 

considered as external dose. Because the INEEL had essentially no separated plutonium, the 

electron dose is the difference between the shallow and deep doses. All nonpenetrating dose for the 

INL should be considered to be electron or beta dose. An alternative method is to use the 

nonpenetrating dose. The number of positive readings for beta exposure is likely to be much smaller 

than for photons, so the missed beta dose will generally be larger. The calculation of missed beta 

Missed Beta Dose
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 38 of 51

dose should 1) determine the measured beta dose and the number of nonzero readings; 2) determine 

the missed dose using MRL/2, the MRLs shown in Table 6-14, and the number of zero readings; then 

3) correct the total dose by multiplying by the factor in the last column of Table 6-10 for the 

appropriate period.

6.5.4

Neutron radiation was present at the INEEL reactors, the 14-MeV neutron generators at RWMC, for a 

short time at TRA and TAN, in small sources used for research at TRA and the INEEL Research 

Center, in calibration laboratories CF 633 and 636, and for calibration of criticality alarms at TRA and 

ICPP. For other locations, there is likely limited missed neutron dose because of the very low 

potential for neutron exposure. To calculate the missed dose, the reconstructor must first determine if 

the person worked near neutrons and which category of neutrons. This can be done best by looking 

for the work location and at whether a coworker or others in the badge reporting group were assigned 

a neutron dose equivalent. The work location code for TRA where the MTR operated is 4 (also 40 

to 45). If no neutron dose was assigned to the worker or coworkers for several months, the dose 

reconstructor should assume that the person was not exposed to neutrons. 

Missed Neutron Dose

If a worker was likely to have been exposed to neutrons, the reconstructor should assign missed 

neutron dose equivalent using Table 6-15 for the times when workers did not have reported neutron 

dose. For the period when NTA film was used, the missed neutron dose should be multiplied by 1.25 

for all facilities except the MTR experimental floor and by 2 for the MTR experimental floor when the 

MTR was operating between 1953 and 1970. Then the dose equivalent is apportioned into the IREP 

groups using Table 6-16. 

Table 6-15. Neutron dosimeter type, period of use, exchange frequency, laboratory 

minimum detectable limit, and maximum annual missed dose. 

Dosimeter Period

Exchange 

frequency

Laboratory 

MRL (mrem)

Maximum 

annual missed 

dose (mrem)

NTA film 1951-1958 Weekly 14 364

NTA film 1959–

September 1976

Weekly 20 520

Biweekly 20 260

Monthly 20 120

TLD October 1976–2006 Biweekly 15 195

Monthly 15 90

Quarterly 15 30

For example, if in 1955 a person was an experimenter at the MTR, and seven of the weekly badges 

recorded a total of 185 mrem neutron dose equivalent, then the missed dose would be 315 mrem 

[(52 - 7) × 14 ÷ 2] so the total dose by the badges would be 500 mrem. Because the badge only sees 

about one-half the MTR neutron dose equivalent (see Section 6.3.4.4), the total dose equivalent is 

1 rem. To convert the 1 rem received from neutrons on the MTR experimental floor to equivalent 

dose, the total dose equivalent is multiplied by the last column of Table 6-16, which results in 200 

mrem to the <10-keV group, 60 mrem to the 10- to 100-keV group, 700 mrem to the 0.1- to 2-MeV 

group, and 280 mrem to the above 2-MeV group for a total equivalent dose of 1.24 rem. 

The neutron missed dose is divided into two historical periods in the following discussion. The first is 

before October 1976 when only NTA film dosimeters were used with supplemental recording of 

thermal neutron doses from B-10 pencil dosimeters. The second period is after September 1976 

when only Hankins albedo dosimeters were used. Table 6-14 summarizes the estimated MRLs for 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 39 of 51

these neutron dosimeters. It is possible to estimate the missed neutron dose using the MRLs 

because the neutron dosimeters were calibrated with neutron sources of energies similar to those 

encountered in the workplace and because most of the neutrons to which workers were normally 

exposed had energies greater than the 500- to 800-keV threshold of the NTA film dosimeters. There 

was, of course, no threshold energy for the measurements using neutron albedo TLD badges.

Table 6-16. Recommended IREP neutron energy fractions and correction factors.

Process Description Operations

Neutron 

energy

Default 

dose 

(%)

Ambient 

dose equiv/ 

dose equiv

Net 

correction 

factor

Instrument 

calibration

Alpha Be source 

calibrations

1951 1993 0.1-2 MeV 20 1.4 0.28

Cf-252 source 

calibrations

1993 2003 2-20 MeV 80 1.1 0.88

Waste 

characterization

RWMC SWEPP 14 MeV 

neutron generator

~1980 2003 0.1-2 MeV 5 1.4 0.07

2-20 MeV 95 1.1 1.05

Neutron source 

based research

2003 0.1-2 MeV 20 1.4 0.28

2-20 MeV 80 1.1 0.88

MTR, ZPPR, and 

TREAT reactors

Experiment floor and 

adjacent rooms during 

operation

1953 1970 < 10 keV 20 1 0.20

10 -100 keV 5 1.1 0.06

0.1-2 MeV 50 1.4 0.7

2-20 MeV 25 1.1 0.28

6.5.4.1 Before October 1976

The use of NTA films for neutron dosimetry before 1976 is documented in various INEEL reports 

(Cusimano 1963; Cipperley 1958, 1968). As noted above, it is possible to estimate the missed dose 

using the MRLs. There are many recorded zeros in the neutron dose data for INEEL workers for two 

reasons: (1) An NTA film was developed and not read per standard criteria, or (2) an NTA film 

indicated a neutron dose equivalent that was less than the film’s 14-mrem MRL. When the MRL for 

NTA film is used to estimate the missed neutron dose, it should be multiplied by 1.25 for most workers 

and by 2 for workers on the MTR experimental floor. 

An estimate of the missed neutron dose in some facilities could also be attainable through use of 

neutron-to-photon dose ratios (NIOSH 2002). However, for the INEEL facilities there are several 

other sources of gamma exposure with no associated neutron exposure, so that approach would be 

erroneous.

6.5.4.2 After September 1976

Since October 1976, the neutron dose has been measured using the Hankins albedo-type TLD. The 

characteristics of this dosimeter are well documented (Gesell et al. 1996), and the MRL to be used in 

estimating missed dose is 15 mrem. A location-specific FNCF has been applied to convert the 

reading to dose equivalent, so no additional adjustments should be required.

For a couple months in the mid-1980s a few office workers without neutron dosimetry were exposed 

to neutrons at TAN (see Section 6.3.4.5). For these individuals, 50 mrem of neutron dose should be 

assigned for each period of exposure. 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 40 of 51

6.6 ORGAN DOSE

Once the Hp(10) adjusted doses have been calculated for each year, the values are used to calculate 

organ doses of interest using the external dose reconstruction guideline (NIOSH 2002). It is 

recommended that the AP geometry should be assumed for the irradiation geometry and for 

conversion to organ dose. The calculated neutron doses in each energy group should be multiplied 

by the conversion factors from ambient dose equivalent to organ dose for AP irradiation from 

Appendix B of NIOSH (2002). For periods before 1981, the conversion factor from exposure to organ 

dose should be used. For 1981 and after, the conversion factor from deep dose equivalent to organ 

dose should be used.

6.7 UNCERTAINTY

Measurement uncertainties arise from many sources. A lognormal distribution should generally be 

assumed unless stated otherwise. For gamma rays, the standards for exposure have existed with 

only minor changes from the 1930s as required for medical uses of radiation. The INEEL used 

ionization chambers standardized by NBS (now the National Institute of Standards and Technology) 

for their calibrations. Use of a phantom for dosimeter irradiation began in the early 1980s, but

backscatter only causes a minor change for high-energy photon dosimetry. The over-response of the 

multielement film badge to deep dose in tissue is due to calibration to exposure, which is somewhat 

greater at low energies than for the deep dose. The INEEL environment did not have a significant 

low-energy photon field such as a plutonium finishing plant, so the nonpenetrating component was 

attributed to beta radiation. A realistic estimate of total uncertainty for photon dosimetry is about 35% 

at 1 sigma. This is roughly consistent with the results in Figure 6-10. For those measurements, the 

standardization instrument contributed some significant uncertainty.

For beta radiations in relation to skin cancer etc, the reported nonpenetrating dose must be divided by 

the fraction of measured dose in Table 6-12. The uncertainty for beta radiation is somewhat larger at 

an estimated 50% at 1 sigma. This is driven by uncertainties in the field geometry and in the fact that 

beta radiation is often stopped by thin materials such as clothing and air. Algorithms are used to 

estimate the dose at a depth of 7 mg/cm2 from dosimeters at depths of 15 to 250 mg/cm2

, and such 

depth differences can change the signal significantly. The difference between a point- and 

planar-source irradiation can confuse an algorithm. Earlier techniques did not provide a thin detector 

with minimal covering, which is important for simulating the skin for beta dosimetry.

For neutron radiations, the situation is more complex. The NTA films used before 1975 did not react 

to low-energy neutrons below 0.5 to 0.8 MeV. Corrections are described for handling this issue. The 

TLD albedo system provides a very indirect way of measuring dose equivalent to a person. Dose to 

workers is primarily due to hydrogen recoils rather than the 6

Li(n,α) reactions. The response of the 

9-in. PNR-4 detector used to standardize the TLD measurements is also due to a different process 

than dose deposition in the human body. The total uncertainty for neutrons is probably a GSD of 

about 1.6. The uncertainty of the base curve for calculating FNCF is a GSD of about 1.35 (Gesell et 

al. 1996) although this uncertainty does not contribute in the DOELAP process, resulting in a total 

GSD of 1.7 for albedo neutron measurements. 

The cause of the greatest uncertainty for neutrons is the variation of dose caused by organ positions 

in the body. For 1-MeV neutrons, the dose facing the source is about a factor of 1,000 higher than the 

dose on the back side of a 30-cm-diameter sphere of tissue-equivalent material. In a work 

environment, the primary direction of the neutrons might be unknown, but it is often from many 

directions, which reduces the impact of this uncertainty driver. For simplicity and because it often is 

true, it is assumed in EEOICPA that the worker irradiation is in an AP geometry (from the front). Note 
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 41 of 51

that in Figure 6-12 the discrepancy in which the dosimeters are reporting about one-half of the 

spectrometer result is because the spectrometer does not simulate the attenuation of the body, so it 

read high by a factor of two.
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 42 of 51

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GLOSSARY

1/E spectrum

For neutrons, the number of neutrons in an energy interval scales as the width of the energy 

interval divided by the energy of the neutrons in that interval.

beta particle

An electron or positron emitted spontaneously at high velocity from the nuclei of certain 

radioactive elements. Most of the direct fission products are (negative) beta emitters. 

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The special unit is the rem, the International System unit is sievert, and 1 sievert equals 

100 rem).

dose equivalent index

Maximum dose equivalent within the ICRU sphere centered at the point in space to which the 

quantity is assigned, HI. The outer 0.07-mm-thick shell is ignored. It is also called the 

unrestricted dose equivalent index.

deep dose equivalent index

Maximum dose equivalent in the ICRU sphere within a core radius of 14 cm. The sphere is 

centered at the point in space to which the quantity is assigned. This quantity is one of the two 

restricted dose equivalent indices.

DOELAP

The DOE Laboratory Accreditation Program (DOELAP) accredits DOE site dosimetry 

programs based on performance testing and onsite reviews performed on a 2-yr cycle. 

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. 

effective dose equivalent, HE 

The weighted average of the dose equivalents in certain organs or tissues of the body, HT, 

each weighted by an organ weighting factor, WT. The organ weighting factors were chosen by 

the ICRP to reflect the relative risk of death from cancer or occurrence of severe hereditary 

effects in the first generations after uniform whole body exposure.

exposure

In the technical sense, the ionization produced by gamma or X-rays in air (roentgen); in the 

generic sense, ionizing radiation applied to matter. 

exposure-to-dose-equivalent conversion factor for photons (Cx) 

The ratio of exposure in air to the dose equivalent at a specified depth in a material of 

specified geometry and composition. The Cx factors are a function of photon energy, material 

geometry (e.g., sphere, slab, or torso), and material composition (e.g., tissue-equivalent 

plastic, soft tissue ignoring trace elements, or soft tissue including trace elements).
Document No. ORAUT-TKBS-0007-6 Revision No. 01 Effective Date: 11/07/2006 Page 51 of 51

linear energy transfer (LET)

The lineal rate of local energy deposition by a charged particle.

minimum reporting level

Based on a policy decision, the minimum dose level that is routinely recorded.

non-penetrating dose

Dose from beta and lower energy photon radiation. Determined from the open window minus 

the shielded.

pencil dosimeters

A type of ionization chamber used by personnel to measure radiation dose. Other names: 

pencil, pocket dosimeter, pocket pencil, pocket ionization chamber.

penetrating dose equivalent

Photon dose measured by shielded INEEL film or elements plus neutron dose equivalent. 

Essentially, personal dose equivalent Hp(10). 

personal dose equivalent, Hp(d)

Radiation quantity recommended for use as the operational quantity to be recorded for 

radiological protection purposes (ICRU 1993). The Personal Dose Equivalent is represented 

by Hp(d), where d identifies the depth (in mm) from the point of reference for dose in tissue. 

For weakly penetrating radiation of significance to skin dose, d = 0.07 mm and is noted as 

Hp(0.07). For penetrating radiation of significance to “whole-body” dose, d = 10 mm and is 

noted as Hp(10). 

polymethyl methacrylate

Scientific name for plastic commonly known as Lucite or Plexiglas.

range

The distance an energetic charged particle will go through a material before it stops. The 

range is an increasing function of energy and depends on the elemental makeup of the 

material and the density.

redacted

To select item(s) to be visible for viewing or for publication by obscuring others.

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.07 mm in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 mm in tissue (sum of penetrating and non-penetrating dose 

equivalent).

tissue rad

Absorbed dose in tissue.